1. Field of the Invention
The present invention relates to methods and systems for separating isotopes from nuclear reactors, and in particular to a method of producing molybdenum-99 (Mo-99) used for medical purposes from the uranyl sulfate nuclear fuel of an aqueous homogeneous solution nuclear reactor.
2. Description of the Prior Art
At the present time more than 50% of the world's annual production of radionuclides are used for medical purposes, such as for the early diagnoses of diseases and for therapy. A basic condition of the use of radionuclides in medicine is the requirement that the radiation exposure of a patient be minimal. This necessitates the use of short-lived radionuclides. A nuclide with a short half-life, however, creates difficulties in transportation and storage. The most used radionuclide for medical purposes is Mo-99 with a half-life of 66 hours. Mo-99 decay results in Tc-99m with a half-life of 6 hours and about 140 keV of gamma (.gamma.) energy convenient for detection. Currently, more than 70% of diagnostic examinations are performed using this radionuclide.
One method of Mo-99 production involves using a target of natural molybdenum or molybdenum enriched in Mo-98 irradiated by a neutron flux in a nuclear reactor. Mo-99 results from a neutron radiation capture .sup.98 Mo(n,.gamma.).sup.99. The irradiated target with Mo-99 then undergoes radiochemical reprocessing. This method, however, has a low productivity and the Mo-99 produced is characterized by a low specific activity due to the presence of Mo-98 in the final product.
Another method of Mo-99 production is based on uranium fission under neutron irradiation of a U-Al alloy or electroplated target in a nuclear reactor. The target contains 93% enriched uranium (U-235). After irradiation, the target is reprocessed by one of the traditional radiochemical methods to extract Mo-99 from the fission products. The specific activity achieved by this method is several tens of kilocuries per gram of molybdenum. A serious disadvantage of this method is the necessity of recovering large amounts of radioactive wastes that are byproducts of the fission process. These wastes exceed the Mo-99 material produced by two orders of magnitude. A 24-hour delay in processing the irradiated uranium targets results in a decrease of total activity by about an order of magnitude, during which time the Mo-99 activity decreases by only 22%. After two days, the activity of the waste byproducts exceeds that of the Mo-99 by a factor of six or seven. The problem of long-lived fission product management is the major disadvantage in the production of Mo-99 by this method.
U.S. Pat. No. 5,596,611 discloses a small, dedicated uranyl nitrate homogeneous reactor for the production of Mo-99 in which the radioactive waste products are recirculated back into the reactor. A portion of the uranyl nitrate solution from the reactor is directly siphoned off and passed through columns of alumina to fix some of the fission products, including Mo-99, to the alumina. The Mo-99 and some fission products on the alumina column are then removed through elution with a hydroxide and the Mo-99 is either precipitated out of the resultant elutriant with alpha-benzoinoxime or passed through other columns. This uranyl nitrate reactor has the advantage of recycling the fission byproducts, but the conventional extraction method to obtain Mo-99 is relatively inefficient.
It is an object of the present invention to produce Mo-99 directly from the uranyl sulfate solution of an aqueous-homogeneous solution nuclear reactor in a manner that minimizes the radioactive byproducts and most efficiently uses the reactor's uranium fuel. The process is relative simple, economical, and waste free.